The positron-emitting 68Ga radionuclide with T1/2=68 min is of enormous practical importance for clinical positron emission tomography (PET). For the generation of radionuclides known radionuclide generators are used, the obtained daughter radionuclides generally having short half-lives T1/2 in comparison to their parent radionuclides.
Radionuclide generators like these are based on a concept of the effective radiochemical separation of decaying parent and daughter radionuclides in such a manner that the daughter nuclide should be obtained in a form with the greatest possible radionuclidic and radiochemical purity.
In comparison to the in-house radionuclide production systems such as accelerators or nuclear reactors, the availability of short-lived radionuclides from radionuclide generators offers an inexpensive and simpler alternative.
The development of radionuclide generators over the past three decades was always marked by the growing range of applications of radionuclides and labelled agents in medicine, especially, for nuclear-medicine diagnostics and therapy. In addition, in recent years, many promising applications of generator-based therapeutic radionuclides in nuclear medicine, oncology and cardiology were developed. This growing importance of radionuclide generators has stimulated a broad development in the production of radionuclides for radionuclide generators, for adequate radiochemical separations as well as for a reliable technical design of radionuclide generator systems. The first generator for applications in the life sciences was already developed in 1920 and via 226Ra (T1/2=1.60·103 a) made available the daughter 222Rn(T1/2=3.825 d) for the production of radon seeds for radiation therapy.
But radionuclide generators did not attain to practical significance until 1951 in the form of the 132Te(T1/2=3.26 d)/132I(T1/2=1.39 h) generators, and to a much more significant extent in 1957 by the pioneering development of the 99Mo/99mTe generators (Stang et al. 1954, 1957). The potential of the daughter nuclide technetium for medical uses quickly became clear and, indeed, the first clinical applications were already described in 1961 which since that time have revolutionized radiopharmaceutical chemistry and nuclear medicine.
The widespread use of the 99Mo/99mTe generator system in nuclear medicine is a typical example of the significance of radionuclide generators for clinks and radiopharmaceutical manufacturers for a broad range of diagnostic radiopharmaceuticals. More than 35,000 diagnostic studies daily with 99mTe involving more than 12 million applications annually are estimated to be conducted just in the USA alone.
Radionuclide generator developments have often ben systematized. Detailed reports about these have devoted themselves to various aspects: parent-daughter half-lives, reactor-produced nuclides, accelerator-produced nuclides, cyclotron production of generator nuclides, ultra-short-life generator-produced radionuclides, generator-based positron-emitting radionuclides, clinical applications.
In the meantime, various other generator systems were developed and some of them have attained to significant practical importance. At present, 68Ge(T1/2=270.8 d/68Ga(T1/2=68 min) generator systems dominate the prior art. Various separation types, 68Ga yields and 68Ge contents are specified below.
The initial generator systems separated 68Ga as an EDTA complex from 68Ge, which was absorbed onto alumina or zirconium oxide, the resulting neutral [68Ga]EDTA solution acting to image tumors. According to an analogous concept 68Ge was retained on antimony oxide Sb2O5 and 68Ga eluated using oxalate solutions. Anion-exchange resins and thinned HF solutions as eluents permitted highly effective separations due to the significant differences of the distribution coefficients of the elements. The 68Ge breakthrough was under 10−4 percent for as many as 600 elutions; the 68Ga yield was greater than 90%.
In all these generator systems a further direct use of the generator eluate for 68Ga labellings was not possible. For this reason, 68Ge/68 Ga generators were developed which led to ionic 68Ga3+ eluates. In these cases 68Ge was fixed onto inorganic matrices such as alumina Al(OH)3 and Fe(OH)3, onto SnO2, ZrO2, TiO2 or CeO2. Tin(IV) oxide SnO2 presented the best parameters in terms of 68Ge breakthrough (10−6-10−5% per bolus) and the 68Ga3+ elution yield (79-80%) in 1 M HCl. Since Ge(IV) is known to form stable complexes with the phenol group, the 68Ge(VI) adsorption onto 1,2,3-trihydroxybenzene(pyrogallol) formaldehyde resins was also exploited. Thus, for a 370 MBq (10 mCi) generator 68Ga3+ elution yields greater than 50% and 68Ge breakthroughs lower than 0.01 ppm were described in the course of the first utilizations.
The 68Ge content defines the radiochemical purity of the separated 68Ga fraction. Even an initial contamination of some 10−2%, corresponding to, for example, 1 μCi 68Ge in a 68Ga fraction of a 10 mCi 68Ge/Ga generator system, appears to be already borderline in connection with a subsequent medical application.
68Ga eluate volumes and chemical purity are other decisive values for the use of 68Ga to synthesize radiopharmaceuticals.
In all the current commercially available 68Ge/Ga generator systems elution volumes of several ml of different HCl solutions are necessary. In addition to considerable volumes, the chemical purity of these 68Ga eluates is an additional critical aspect of 68Ge/Ga generator systems.
The highest chemical purities, especially a minimum content of diverse metallic cations, are necessary for efficient labelling reaction with high yields. This applies especially in cases where labelling chemistry is conceived making use of bifunctional chelators. In this context, even small amounts of stable 68Zn as a direct decay product of 68Ga, of titan, in cases whether the 68Ge/Ga generator system ion exchanger column is made of TiO2, and especially also of iron, can prevent high labelling yields.
At present, commercially available 68Ge/Ga generator systems are limited to effective 68Ga elutions and do not comprise modalities for volume minimizing and purification of the generator eluates or labellings of potential radiopharmaceuticals.
Initial volumes of the eluates amount to from a few ml to 10 ml of HCl solutions of various concentrations. Both the labelling reactions as well as the filling of balloons generally require smaller volumes of roughly 0.5 ml to 0.1 ml. Hence, chemical or technological strategies are necessary which immediately afterwards the initial generator elution lower the eluate volume.
Secondly, the generator eluate can contain chemical and radiochemical contaminations which prevent the efficient exploitation of radiochemical labels with high yield. These chemical contaminations can come from:                the generator column material (for example TiO2);        trivalent Fe, which is ubiquitous in traces and can especially be introduce with diverse electrolytes during the manufacture or use of the generator;        68Zn as a stable metallic contamination, which system-inherent is continuously generated as a decay product of 68Ga on the generator column;        68Ge as parent radionuclide, whereby even slight contaminations of less than 0.01% of the 68Ge in the eluate represent a similar number of atoms as the 68Ga itself,        and which can act in a radiotoxic and chemotoxic manner.        
Besides the aspect of the chemical contaminations by 68Ge in the generator eluate, this contamination is also radiochemically relevant, especially in view of the potential medical applications. Hence, the post-elution procedure should explicitly comprise a chemical strategy for the further separation of the 68Ge.
Thirdly the 68Ga labelling of potential radiopharmaceutical assumes a central role for which the corresponding chemical reaction parameters must be optimized.
The trivalent Gallium hydrolyzes already from pH>2 and has a marked tendency to absorb on the surfaces of glass and polymers at pH>3, especially in the condition of the low 68Ga concentrations (no-carrier-added), as they arise from the generator system. Finally, special reaction conditions must be chosen in the case of the labelling chemistry of targeting vectors using bifunctional chelators such as DOTA due to the complexing kinetics as well as due to the aqueous chemistry of the Ga(III) cation.
In addition to the metallic contaminations associated with the operation of the generator system which are eluated along with the 68Ga, the contaminations contained in the buffer systems generally used for 68Ga labels can in some cases also handicap high labelling yields.
Processes related to solvent vaporization for the reduction of volumes of generator eluates or of the final solutions of the 68Ga radiopharmaceutical an also lead to losses of activity both owing to the related longer duration of the process as well as owing to adsorption losses along the vessel walls.
Individual experimental conceptions towards the minimization of the eluate volume for 68Ge/Ga generator systems have been developed. Some of these realizations (Meyer et al. 2004, Velikyan et al. 2004) minimize the initial eluate volumes by mixing with several ml of concentrated HCl, whereby a total of 6 M HCl solution of an increased volume of approximately 15 ml results. This large volume is then transferred to an anion-exchanger column on which the 68Ga is adsorbed. Then the 68Ga is eluated with less than 1 ml water. Although this time-consuming procedure does realize a reduction in the volume of the 68Ga fraction, there is no obvious parallel strategy for separating chemical contaminations out of the initial generator eluate.
Subsequently, 10-20 nmol DOTATOC is then added to this fraction in a small volume of an aqueous 1 M HEPES or other buffer solution. Here too, potential contaminations by the concentrated buffer system cannot be excluded.
These factors can be the reason for the fact that the labelling yields of 68Ga-DOTATOC and analogous compounds achieved under standard heating protocol have been merely 58±20% (Meyer et al. 2004). Greater yields are described by microwave-supported heating (Velikyan et al. 2004).
Overall, none of the currently commercially available generator systems comprise the corresponding chemical or technological strategies to solve the problems mentioned as a whole.
The problem of the invention is to create a method and a device which makes available highly pure 68Ga eluate which is largely free of chemical and radiochemical contaminations with a high yield and very low eluate volume. In the framework of this problem the chemical reaction parameters such as the pH value of the 68Ga for the labelling of the labelling precursors should also be optimized. Furthermore, a procedure for labelling potential radiopharmaceuticals for positron emission tomography should be provided.
At the same time, all the process steps should meet the requirements of simple and routine use in a medical environment.